5.3. ORIGEN Reactor Libraries
B. Hiscox, B. R. Betzler, B. J. Ade
5.3.1. Description of ORIGEN Reactor Libraries in SCALE
The ARP code creates burnup-dependent ORIGEN cross-section libraries by interpolating over reactor cross-section libraries generated in advance using reactor physics transport methods. The reactor cross-section libraries distributed in SCALE include many classes of commercial power reactor designs and a range of fuel assembly designs.
Cross-section libraries suitable for use with ORIGEN are available in SCALE for the following reactor and fuel assembly designs, summarized in Table 5.3.1.
BWR 7 \(\times\) 7, 8 \(\times\) 8-1, 8 \(\times\) 8-2, 9 \(\times\) 9-2, 9 \(\times\) 9-9, 10 \(\times\) 10-9, 10 \(\times\) 10-8, SVEA-64, SVEA-96, and SVEA-100;
PWR 14 \(\times\) 14, 15 \(\times\) 15, 16 \(\times\) 16, 17 \(\times\) 17, 18 \(\times\) 18;
CANDU reactor (19-, 28-, and 37-element bundle designs);
Magnox graphite reactor;
Advanced Gas-Cooled Reactor (AGR);
VVER 440 and VVER 1000;
RBMK;
IRT;
MOX BWR 7 \(\times\) 7, 8 \(\times\) 8-1, 8 \(\times\) 8-2, 9 \(\times\) 9-2, 9 \(\times\) 9-9, 10 \(\times\) 10-9, 10 \(\times\) 10-8, SVEA-64, SVEA-96, and SVEA-100;
MOX PWR 14 \(\times\) 14, 15 \(\times\) 15, 16 \(\times\) 16, 17 \(\times\) 17, 18 \(\times\) 18.
All of the libraries distributed with SCALE were developed using ENDF/B-VII.1-based 252-group cross section libraries and the SCALE 6.3 TRITON module. The assembly design and the identifier for each library are summarized in Table 5.3.2. All TRITON templates used to generate these libraries are provided with the release; these templates have been developed to reflect the most recent publicly available data and to adhere to current SCALE best practices. MOX templates use the same set of BWR and PWR lattices, with MOX compositions replacing the UO:sub:2 fuel.
Additionally, a Python script (slig.py) has been developed to semi-automate the
process of generating sets of libraries from information in these templates.
This script calls another Python script to automatically generate a PDF
documentation file with LATEX. This SLIG script is made available to the user;
more details are available in the SLIG manual.
Reactor type |
Assembly Design Description |
Library Name(s) |
PWR LEU and MOX |
Babcock & Wilcox 15 \(\times\) 15 |
bw15x15, mox_bw15x15 |
Siemens 14 \(\times\) 14 |
s14x14, mox_s14x14 |
|
Siemens 18 \(\times\) 18 |
s18x18, mox_s18x18 |
|
Westinghouse 14 \(\times\) 14 |
w14x14, mox_w14x14 |
|
Westinghouse 15 \(\times\) 15 |
w15x15, mox_w15x15 |
|
Westinghouse 17 \(\times\) 17 |
w17x17, mox_w17x17 |
|
Westinghouse 17 \(\times\) 17-OFA |
w17x17_ofa, mox_w17x17_ofa |
|
Westinghouse CE 14 \(\times\) 14 |
ce14x14, mox_ce14x14 |
|
Westinghouse CE 16 \(\times\) 16 |
ce16x16, mox_ce16x16 |
|
BWR LEU and MOX |
ABB 8 \(\times\) 8-1 |
abb8x8-1, mox_abb8x8-1 |
ATRIUM 9 \(\times\) 9-9 |
atrium9x9-9, mox_atrium9x9-9 |
|
ATRIUM 10 \(\times\) 10-9 |
atrium10x10-9, mox_atrium10x10-9 |
|
General Electric 7 \(\times\) 7-0 |
ge7x7-0, mox_ge7x7-0 |
|
General Electric 8 \(\times\) 8-1 |
ge8x8-1, mox_ge8x8-1 |
|
General Electric 8 \(\times\) 8-2 |
ge8x8-2, mox_ge8x8-2 |
|
General Electric 9 \(\times\) 9-2 |
ge9x9-2, mox_ge9x9-2 |
|
General Electric 10 \(\times\) 10-8 |
ge10x10-8, mox_ge10x10-8 |
|
SVEA 64(8 \(\times\) 8-1) |
svea64-1, mox_svea64-1 |
|
SVEA 96(10 \(\times\) 10-4) |
svea96-0, mox_svea96-0 |
|
SVEA 100(10 \(\times\) 10-0) |
svea100-0, mox_svea100-0 |
|
VVER |
VVER-440 flat enrichment |
vver440 |
VVER-440 radial enr. profile, avg. 3.82% |
vver440_3.82 |
|
VVER-440 radial enr. profile, avg. 4.25% |
vver440_4.25 |
|
VVER-440 radial enr. profile, avg. 4.38% |
vver440_4.38 |
|
VVER-1000 |
vver1000 |
|
RBMK |
RBMK 1000, flat enrichment |
rbmk1000 |
CANDU |
19-element bundle design |
candu19 |
28-element bundle design |
candu28 |
|
37-element bundle design |
candu37 |
|
Magnox |
Magnox graphite reactor (Calder Hall) |
magnox |
AGR |
Advanced gas-cooled reactor |
agr |
IRT |
IRT-2M (3 tube) |
irt2m3tube36enrich, irt2m3tube |
IRT-2M (4 tube) |
irt2m4tube36enrich, irt2m4tube |
|
IRT-3M (6 tube) |
irt3m6tube36enrich, irt3m6tube90enrich |
|
IRT-3M (8 tube) |
irt3m8tube36enrich, irt3m8tube90enrich |
|
IRT-4M (6 tube) |
irt4m6tube |
|
IRT-4M (8 tube) |
irt4m8tube |
|
LEU = Low-enriched uranium MOX = Mixed-oxide |
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The libraries for BWR and RBMK assembly designs all include variable coolant density. Table 2 lists the number of cross-section files associated with each individual assembly design; this number is often a product of the number of enrichments (or Pu contents for MOX assemblies) and the number of coolant density values. Unless otherwise noted in Table 5.3.1, the PWR and BWR LEU libraries were generated for ten enrichment values: 0.5, 1.5, 2.0, 3.0, 4.0, 5.0, 6.0, 7,0, 8.0, and 8.5 wt% 235U. The maximum enrichment and burnups of most of the BWR and PWR libraries were extended in SCALE 6.3 from their historical maximum values of 6.0 wt% enrichment and 70.5 GWd/MTIHM burnup to 8.5 wt% and 82.5 GWd/MTIHM, respectively.
The cross sections are represented at several burnup values in the range from 0 to the maximum burnup listed. The number of burnup values on the library has been optimized with a post-processing script to minimize the size of the data libraries without losing significant fidelity in the cross section data. Note that these optimized burnup steps are not the same burnup steps used in the TRITON calculations; more burnup steps are used to calculate the depletion of the fuel. The fuel assembly designations used for many of the BWR designs include the basic lattice type, followed optionally by the number of empty (non fuel) lattice positions. For example, the 9 \(\times\) 9-9 designation refers to the Atrium design with a 9 \(\times\) 9 assembly lattice with 9 non-fuel (water) locations. Additional data on the reactor libraries is available in the library information PDF file released with SCALE.
This document includes details on the fuel lattice type, the fuel vendor (where design information is specific to a vendor design), the assembly model, basic fuel design and reactor operating data, and the range of the variable parameters associated with the libraries (e.g., enrichment range, burnup range, etc.). This document references separate reports from which a more detailed description of the data for these libraries can be found.
Assembly type |
Enrichments [ wt. %235U ] |
Coolant/moderator densities [g/cm3] |
Maximum burnup [GWd/MTIHM] |
Number of Libraries |
PWR LEU |
0.5, 1.5, 2, 3, 4, 5, 6, 7, 8, 8.5 |
~0.73 |
82.5 |
10 |
BWR LEU |
0.5, 1.5, 2, 3, 4, 5, 6, 7, 8, 8.5 |
0.1, 0.3, 0.5, 0.7, 0.9 |
82.5 |
50 |
PWR MOX |
* |
~0.73 |
82.5 |
15 |
BWR MOX |
0.1, 0.3, 0.5, 0.7, 0.9 |
82.5 |
75 |
|
AGR |
0.5, 1.5, 2, 3, 4, 5 |
1.65 |
48.75 |
6 |
CANDU |
0.711 |
0.8445 |
13.725 |
3 |
IRT |
19.75, 36, 80, 90 |
0.989 |
159.3 |
12 |
Magnox |
0.7, 0.8, 0.9, 1 |
1.628 |
13.725 |
4 |
RBMK |
1.8, 2.2, 2.6, 3 |
0.15, 0.28, 0.41, 0.54, 0.67, 0.8 |
24.375 |
24 |
VVER-440 |
1.6, 2.4, 3.6, profiled |
0.73 |
82.5 |
6 |
VVER-1000 |
0.5, 1.5, 2, 3, 4 5, 6, 7, 8, 8.5 |
0.7145 |
82.5 |
10 |
* Pu contents [% heavy metal]: 4, 7, 10; 239Pu contents [% of Pu]: 50, 55, 60, 65, 70 |
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5.3.2. How to Generate ORIGEN Cross-Section Libraries
This section describes the basic procedures one needs to follow to generate ORIGEN cross-section libraries for fuel types not included in SCALE. Note that the extension of the methods and data beyond 5 wt% enrichment has only been recently investigated by ORNL [ORIGEN-RX-LIBSCSHW21, ORIGEN-RX-LIBSHCWS21]. While no unexpected or anomalous trends were observed in trends beyond the limits of previous reactor data libraries (i.e., 5.0 wt% enrichment and 60 GWd/MTIHM), there is nonetheless a current lack of accessible validation data for these higher-burnup domains. In general, effects of higher enrichment were found to offset higher burnup limits. Users are encouraged to carefully test and validate applications of these libraries in these extended domains.
The following discussion and examples relate primarily to generating commercial LWR fuel libraries. Note that the parameter ranges used as examples are not necessarily appropriate to non-LWR applications, and will need to be modified for different reactor types and fuel designs.
The first step is to construct a physics model of the fuel lattice with the descriptions of the reactor assembly under consideration. For a given initial fuel enrichment, a TRITON depletion calculation is performed using one of the depletion analysis sequences of SCALE. TRITON can use either an explicit 2-D representation of the fuel assembly using the NEWT discrete ordinates transport code, or a 3-D representation of the assembly using the KENO Monte Carlo transport code. TRITON allows multiple fuel types to be defined and depleted independently.
The depletion analysis sequences are used to simulate irradiation and depletion of the fuel over the required irradiation history. A burnup analysis is typically modeled using a series of time intervals, or cycles. During the simulation, cross-sections that are representative of the mid-point of each cycle are created and saved in the library by the depletion sequence. In addition, cross sections that are representative of fresh fuel conditions (zero burnup) are now automatically saved in the library, without the need for the user to add a small burnup step to represent initial cross sections, as required in earlier versions of SCALE.
Past experience in creating LWR fuel libraries has indicated that depletion simulations with burnup steps of 3,000 MWd/MTU are generally adequate to represent the cross section variations with burnup. Each set of burnup-dependent cross sections is stored within the single library, and each set is accessed sequentially by its position in the library. The position 1 set contains fresh-fuel cross sections and the remaining cross-section sets (positions) correspond to burnup levels characterizing the midpoint of each burnup step in the depletion sequence calculation.
For fuels with multiple enrichment values, the above procedure is repeated for each enrichment. For fuels with a single enrichment value (e.g., natural uranium), only one depletion case is required. Cross-section changes with enrichment are generally represented using approximately 1.0 wt % increments, e.g., 0.5, 1.5, 2.0, 3.0, 4.0, 5.0, 6.0, 7.0, 8.0, and 8.5 wt % of 235U. Because the cross-section variation is generally well behaved and smooth with changes in enrichment, it may be possible to maintain accuracy using fewer enrichment points. For BWR designs, or reactor designs involving considerable moderator density variations, one has also to consider the effect of water density variation on the cross sections. The recommended water density values for BWRs are typically 0.1 to 0.9 g/cm3, with increments of about 0.2 g/cm3. Again, fewer points may be adequate for many applications. Water density variation has typically not been included for PWR libraries developed at ORNL because of the relatively small variation in pressurized water reactors. When the variation in moderator density is considered, the libraries must be calculated for each combination of enrichment and moderator density. This may involve a considerable number of depletion analysis simulations to prepare all of the required cross-section libraries.
The SLIG utility has been created to help automate the creation of input files that represent each of the discrete parameter combinations. Instead of having to manually create input files for each parameter value, a single template input file is constructed with generic flags substituted for the parameters that vary from case-to-case (i.e., enrichment and water density). The specific values required for each parameter are specified in the header of the template file; SLIG reads the template and substitutes the values for the generic flags in the input file. SLIG generates a unique subdirectory and input file for each parameter combination. A description of SLIG and examples are provided in the SLIG manual.
Another utility provides an optional procedure that can be used to reduce the size of the cross-section libraries by eliminating some of the burnup-dependent cross-section sets in the libraries in regimes where the cross sections do not vary appreciably with burnup. In general, the change in cross-section values is more pronounced early in irradiation, and tends to approach asymptotic values at higher burnup. Therefore, interpolation of the cross sections using fewer data points may yield acceptable accuracy in this range. This utility is currently unavailable in SCALE 6.3.
As an example of this procedure, the basic steps in creating a set of cross-section libraries for a typical PWR assembly are illustrated. In this example the burnup is calculated up to 72 GWd/MTU. To generate these basic cross-section libraries, the following steps are performed for each fuel enrichment value:
1. Create a 2-D or 3-D reactor physics model using TRITON for the fuel assembly design and fuel type being considered.
2. Run a depletion analysis calculation to the maximum burnup under consideration. For example, set a maximum burnup at 72 GWd/MTU, and simulate using 24 intervals (cycles) with one library per cycle. Each of these cycles is 3 GWd/MTU in size. The library generated by this analysis will contain 25 sets of cross sections: initial fresh fuel cross sections plus 24 burnup-dependent cross sections.
3. Include as many extra nuclides in the fresh fuel composition specification as is practical in trace concentrations (e.g., 1E-20 atoms/barn-cm). This procedure ensures that updated cross sections for each nuclide added will be used in creating the ORIGEN library. The depletion analysis sequences automatically add many of the most important actinides and fission products. However, the addition of more nuclides by the user will result in more complete updating of the cross-section data. The addition of trace nuclides is handled automatically in TRITON using the ADDNUX= parameter entry.
4. The above procedure is repeated for each enrichment value. If moderator density variation is to be included, a separate calculation is required for each enrichment-moderator density value combination.
5. Finally, the ARPDATA.TXT file must be updated to include the data for the new cross-section libraries created. This file contains information on the library, including the number of values for each variable parameter, the parameter values, the burnup values for each library position, and the actual library file names. The burnup values for each set of cross sections in the library are printed in the TRITON output file to assist the user in determining the appropriate library entries. Once the library information is registered in the ARPDATA.TXT file, it can be used directly by ORIGEN.
References
- ORIGEN-RX-LIBSCSHW21
Riley M. Cumberland, Ryan T. Sweet, Ugur Mertyurek Robert Hall, and William A. Wieselquist. Isotopic and Fuel Lattice Parameter Trends in Extended Enrichment and Higher Burnup LWR Fuel, Volume II: BWR. Technical Report ORNL/TM-2020/1835, Oak Ridge National Laboratory, Oak Ridge, TN (USA), March 2021. doi:10.2172/1782042.
- ORIGEN-RX-LIBSHCWS21
Bob Hall, Riley Cumberland, William Wieselquist, and Ryan Sweet. Isotopic and Fuel Lattice Parameter Trends in Extended Enrichment and Higher Burnup LWR Fuel Vol I: PWR fuel. Technical Report ORNL/TM-2020/1833, Oak Ridge National Laboratory, Oak Ridge, TN (USA), February 2021. doi:10.2172/1779134.